This invention relates to fluid cooled nuclear reactors and means for minimizing the effect of loss of coolant through a cold leg break.
In the event of an accident in which there is a break in the cold leg of the reactor coolant system, it has been postulated that the entire coolant medium which absorbs and removes the heat generated in the nuclear core will be lost or at least considerably decreased. Although control elements are inserted into the core to terminate the core's fission process upon the occurrence of such an accident, decay heat generated by the already formed fission products is capable of causing fuel or clad melting if sufficient emergency cooling of the fuel is not immediately accomplished. Furthermore, the overheating of the fuel cladding can result in a severe adverse chemical reaction in the reactor core, which may not be reversed by later cooling procedures. Accordingly, it is necessary to provide a sufficient coolant flow immediately after the accident to ensure that this heating of fuel and cladding is kept within acceptable limits.
Many prior art systems have been developed to inject emergency core coolant water into the reactor pressure vessel when such an accident occurs. If the emergency core coolant water does not reach the reactor core within the first minute following the accident, the entire reactor core, fuel and supporting structure begins to melt down and slump to the bottom of the reactor vessel. Emergency cooling water injected at this stage may well amplify the disaster as the now molten metals can react violently with water, generating large quantities of heat, releasing steam and hydrogen in amounts and at pressures that can themselves breach the containment. If the containment vessels themselves do not burst, it has been postulated that the molten mass of fuel would continue to melt downward, fed by the heat generated by fission product radioactivity.
It is the function of the emergency core cooling system to resupply the core within this first minute with coolant water in the event of the loss of primary coolant in a loss of coolant accident. The emergency core coolant water is normally injected into the inlet nozzles of the reactor and allowed to flow downwardly along the normal coolant path to the bottom of the vessel and then upwardly to the reactor core. A possible difficulty with this method of injecting emergency core coolant into the reactor is that large quantities of steam are generated by the hot core which creates a pressure buildup in the core and the reactor outlet plenum. This buildup of pressure not only impedes the further coolant from reaching the core but also may displace the water already in the core downwardly out of the core. This excess pressure also creates a further problem in that in the ordinary nuclear steam supply system, the pressure buildup has only two paths for its relief. If a break occurs in the cold leg between the reactor vessel and the coolant pump causing the loss of coolant accident, then the first path for pressure relief is down through the middle of the reactor core and up through the outer annulus between the core support barrel and the pressure vessel walls and out through the cold leg to the break. This flow is opposite to the flow of the emergency core cooling water being injected into that leg thereby impeding the flooding of the reactor core from the bottom. The second path for pressure relief is from the plenum above the reactor core out through the hot leg to the steam generator, through the steam generator and through the coolant pump to the cold leg break where the steam is discharged to the atmosphere. The steam that flows through this path is being driven by the pressure differential between the high pressure in the reactor plenum above the core and atmospheric pressure. Consequently, the steam flow rate is very high and the steam tends to drive the reactor pump like a turbine. This circumstance produces the substantial danger of over speeding the reactor coolant pump to the point where the massive fly-wheel connected to the reactor coolant pump shaft disintegrates and eventually would cause severe damage to the surrounding equipment. It is to the solution of these problems that the present invention is directed.
One prior art solution to these problems is disclosed in the April, 1970 issue of "Power Magazine" on pages 90-91. This prior art solution consists of the incorporation of a plurality of swing check valves or flapper valves in the core support shield between the plenum over the reactor core and the outer annulus between the core support shield and the reactor pressure vessel. These valves operate to equalize the pressure differential between the reactor's upper plenum and the annulus by swinging open on the occurrence of a very low differential pressure and thereby allowing the generated steam to flow directly from the reactor to the annulus and out through the cold leg rupture. These valves, however, have the difficulty that they are mounted on the interior of the reactor pressure vessel and consequently become highly radioactive. They also have the further difficulty of being somewhat complicated which increases the possibility of failure. Failure of one of these valves would substantially impair the operation of the nuclear reactor power system. Furthermore, if a physical breakdown of one of these valves were to occur, the failure may cause a loose part to be released to the incoming coolant water flow. The loose part could then possibly be swept into a flow blocking position which could cause unnecessary overheating of the reactor fuel elements, or would turbulently clatter around in the lower plenum, destroying or severely damaging critical elements that may be positioned in the lower plenum.